Journal of Nuclear Fuel Cycle and Waste Technology

This study examines the thermal relaxation and pressure attenuation behavior of Bentonil-WRK, a Ca-type bentonite, under a constant temperature of 130°C. A one-month pressure-measurement experiment was conducted using a stainless-steel cell equipped with a high-temperature load cell and a pressure–volume controller, applying hydration and thermal loading simultaneously. The specimen was compacted to a dry density of 1.60 g·cm^-3. Following initial heating and hydration, the measured total pressure reached approximately 3,700 kPa, after which it exhibited gradual, non-linear attenuation throughout the 30-day period. Five empirical attenuation models were evaluated, and the logarithmic model provided the best fit (R² = 0.996; RMSE = 6.82 kPa), capturing both the early attenuation and slow long-tail behavior. Using this model, the pressure trend was extrapolated to 1,000 years under a conservative assumption that thermal relaxation continues without convergence at 130°C. Because this study was limited to single-temperature, dry density, and hydration condition, the extrapolated results should be interpreted as a methodological demonstration rather than a prediction of repository performance. Nevertheless, the findings provide useful insights into the empirical characterization of thermal relaxation in Ca bentonites under elevated temperatures.

As nuclear power plants approach the end of their spent fuel storage capacity or are scheduled to be decommissioned, the need for dry storage is increasing. In preparation for dry storage, spent fuel must be packaged to maintain mechanical integrity during handling and transportation in accordance with the legal requirements for delivery of domestic spent fuel dry storage (Spent Fuel Delivery Regulations, Nuclear Safety and Security Commission Notice), and integrity must be maintained to prevent gross rupture of spent fuel rods. Intact spent fuel that has not been damaged may satisfy the legal requirements for dry storage, but the structural integrity of damaged spent fuel must be confirmed for handling and transportation. Various types of damage may occur during handling of spent fuel in a nuclear power plant and based on the damage to the grids supporting the fuel rods reported by the IAEA, an integrity assessment was performed by assuming damage to the outer surface of spacer grids supporting fuel rods. When damage such as loss of the outer plate occurs in some of the grids of the fuel assembly, buckling of the surrounding normal grids that are not damaged and excessive deformation of the fuel rods may occur during transportation. In this study, we evaluated the grid buckling and fuel rod bending to assess the mechanical integrity during handling and transportation of spent fuel by assuming grid damage.

Understanding how saline water intrusion alters the performance of bentonite buffer is critical for the reliable safety assessment of a high-level radioactive waste repository. This study investigated sorption thermodynamics of Cs(I) and Sr(II) onto Na-exchanged Bentonil-WRK montmorillonite. Batch sorption experiments were conducted under ambient conditions (T = 25℃, pH 4–10, S/L = 5 g·L⁻¹, I = 0.01 mol·L⁻¹ NaCl), followed by inverse modeling of the resulting data. The sorption of Cs(I) was dominated by cation exchange (XNa + Cs⁺ ⇌ XCs + Na⁺, log K = 1.48 ± 0.02) with a minor contribution from edge surface complexation at silanol site (≡SiOH + Cs⁺ ⇌ ≡SiOCs + H⁺, log K = −5.36 ± 0.70). In contrast, Sr(II) uptake involved a more complex mechanism comprising cation exchange (2XNa + Sr²⁺ ⇌ X2Sr + 2Na⁺, log K = 0.47 ± 0.06), interlayer ion-pairing with chloride (XNa + Sr²⁺ + Cl− ⇌ XSrCl + Na⁺, log K = 3.05 ± 0.33), and edge complexation (≡SiOH + Sr²⁺ ⇌ ≡SiOSr⁺ + H⁺, log K = −6.04 ± 0.94). The derived thermodynamic parameters complement the radionuclide sorption database for Ca-type bentonite, reducing uncertainty by enabling process-based evaluation of clay buffer performance under long-term geochemical evolution scenarios.

In this study, ORIGEN-2 one-group cross-section libraries were developed for prism-type VHTR whole-core models to enable accurate and efficient depletion and decay heat calculations. These libraries were generated through continuous-energy Monte Carlo depletion simulations using the McCARD code. Two computational approaches were compared. The Core Depletion (CD) approach - previously developed and used as a reference - performs whole-core depletion with McCARD, followed by decay heat calculation using ORIGEN-2. In contrast, the Point Depletion (PD) approach applies McCARD-generated one-group cross-sections within ORIGEN-2 for both depletion and decay heat analysis. Among the library-generation methods evaluated, the batch-wise library - constructed via localized tallying from whole-core depletion - yielded nuclide inventories and decay heat results in close agreement with those of the CD approach. Its improved accuracy stems from its ability to capture the soft neutron spectrum induced by large graphite reflectors and the detailed burnup history - features not properly represented in single-block models or in whole-core depletion with unified tallying. Notably, the batch-wise method enabled precise prediction of long-term actinide decay heat and achieved high accuracy in estimating major source-term nuclide inventories, with relative errors remaining below 5% in most cases when compared to the reference solution, offering a validated and efficient framework for VHTR core analysis, safety evaluation, and spent fuel management.